Thermal hydraulics and mechanics deals with the physics and mechanics of the flow and energetic transfer of liquids, and its interactions with the structures around them in large complex systems, such as nuclear reactors.
Systems thermal-hydraulic codes have dominated flow modelling for nuclear reactor systems analysis. Single-phase computational fluid dynamics (CFD) methods have a long history, beginning with special codes mainly developed at government laboratories, and expanding rapidly after widespread acceptance of commercial and open source CFD codes. As CFD methods become more widespread, coupling these methods to system codes, for both traditional light water reactors (LWRs) and next generation systems is becoming increasingly a domain for scientific developments.
Created in 2019 under the guidance of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Expert Group on Reactor Core Thermal-Hydraulics and Mechanics (EGTHM) performs specific tasks associated with core thermal-hydraulics aspects of present and future nuclear power systems.
EGTHM provides expert advice to the WPRS and the nuclear community on the development needs (data and methods, validation experiments, scenario studies) for multi-scale core thermal-hydraulics modelling and simulation of existing and proposed nuclear reactor systems. A key activity associated with this objective is the identification and preservation of appropriate experimental data. The expert group provides member countries with the guidance and processes for certifying experimental data for its use as a stand-alone core thermal-hydraulic validation or for uses as part of validation pyramid of multi-physics modelling and simulation tools.
The group also monitors, steers and supports the continued development of the The International Experimental Thermal Hydraulics Systems (TIETHYS) database. They also facilitate the dissemination of technical information and knowledge through activities such as workshops, benchmark studies and training activities. The group provides state-of-the-art best estimate and uncertainty analysis for many types of reactor systems.
Additionally, the group publishes expert guidance in the domain of radiation transport and shielding. The work conducted within EGTHM has led to numerous conference and journal publications, in addition to the NEA technical reports published.
The Benchmark for Uncertainty Analysis in Best-Estimate Modelling (UAM) for Design, Operation and Safety Analysis of Light Water Reactors (LWRs) is an international high-visibility benchmark for uncertainty analysis in best-estimate coupled code calculations for design, operation, and safety analysis of LWRs. The annual workshops are attended by many experts in industry, research institutes, national laboratories, academia, and government agencies.
The Boiling Water Reactor Turbine Trip (BWRTT) Benchmark was established to challenge the coupled system thermal-hydraulic/neutron kinetics codes against a Peach-Bottom-2 (a GE-designed BWR/4) turbine trip transient with a sudden closure of the turbine stop valve.
The expert group provides advice to the Working Party on Scientific Issues of Reactor Systems (WPRS) and the nuclear community on the scientific development needs (data and methods, validation experiments, scenario studies) of sensitivity and uncertainty methodology for modelling of different reactor systems and scenarios.
The overall objective of the VVER-1000 Coolant Transient (V1000CT) Benchmark was to assess computer codes used in the safety analysis of water-water energetic reactor (VVER) power plants, specifically for their use in reactivity transients in VVER-1000.